Studies of the steam generator degraded tubes behavior on BRUTUS test loop

Conference · Tue Apr 01 00:00:00 EST 1997

OSTI ID:489362

Chedeau, C; Rassineux, B ^[1]; Flesch, B ^[2]

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1. EDF/DER/MTC, Moret Sur Loing (France)
2. EDF/EPN/DMAINT, Paris (France); and others

Studies for the evaluation of steam generator tube bundle cracks in PWR power plants are described. Global tests of crack leak rates and numerical calculations of crack opening area are discussed in some detail. A brief overview of thermohydraulic studies and the development of a mechanical probabilistic design code is also given. The COMPROMIS computer code was used in the studies to quantify the influence of in-service inspections and maintenance work on the risk of a steam generator tube rupture.

• Research Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Engineering Technology; Electricite de France (EDF), 69 - Villeurbanne (France); Battelle Memorial Inst., Columbus, OH (United States)
• OSTI ID: 489362
• Report Number(s): NUREG/CP-0155; CONF-9510432-; ON: TI97004806; TRN: 97:011503
• Resource Relation: Conference: Seminar on leak before break in reactor piping and vessels, Lyon (France), 9-11 Oct 1995; Other Information: PBD: Apr 1997; Related Information: Is Part Of Proceedings of the seminar on leak before break in reactor piping and vessels; Faidy, C. [ed.] [Electricite de France, Villeurbanne (France)]; Gilles, P. [ed.] [Framatome, Paris (France)]; PB: 773 p.
• Country of Publication: United States
• Language: English

Related Subjects

22 NUCLEAR REACTOR TECHNOLOGY
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS
STEAM GENERATORS
PRESSURE TUBES
RUPTURES
STRESS CORROSION
INCONEL 600
C CODES
FLOW RATE
DIMENSIONS
PWR TYPE REACTORS